The present invention relates to a method and apparatus for processing aqueous, radioactive wastes for noncontaminating and safe handling, transporting and final storage in which aqueous radioactive waste solutions containing nitric acid and/or nitrates are continuously denitrated with formic acid, and are spray-dried and calcinated. The resulting calcinate is mixed with glass former substances, the mixture is melted and the melt is caused to solidify into a glass, glass ceramic or glass ceramic-like block, and the waste gases produced during denitration, drying and calcination are conducted through a filter system in order to remove solid particles that have been carried along with the gas.
For safe handling, transport and storage of radioactive wastes, particularly if they are to be stored over long periods of time, only those solidification products can be used which have high chemical, mechanical and radiolytic stability. Solidification products containing highly radioactive wastes must also have a high thermal stability. A suitable solidification matrix for such wastes has been found to be borosilicate glasses which are also encountered in nature at an age up to 10.sup.5 years. These glasses are capable of absorbing large amounts of fission product oxides and corrosion products from the wastes with a simultaneous relatively great insensitivity with respect to the particular composition of the fission product oxides and corrosion products.
During the melting process for solidification of the wastes, complete homogenization must take place during their stay in the melting crucible. Since sufficient stability of the material from which the metallic melting crucible is made is assured only up to about 1200.degree. C., this temperature constitutes an upper limit for the temperature of the solidification melt. On the other hand, a viscosity of less than 100 poise is required. This requirement is a result of the configuration of the melt outlet so that the flow of glass can be interrupted by cooling.
The softening point (10.sup.8 poise) of glass solidification products must lie, for reasons of later storage, for example, storage in rock salt, above 700.degree. C. Experimental melts using simulated, i.e., inactive, fission product oxide mixtures have shown that, for the incorporation of radioactive fission product oxide mixtures and other solid mixtures of radioactive wastes in quantities up to 25 percent by weight of the solidification product, a basic glass type, having a composition, in percent by weight of the basic glass, of 52.5% SiO.sub.2, 10.0% TiO.sub.2, 2.5% Al.sub.2 O.sub.3, 10.0% B.sub.2 O.sub.3, 5.0% CaO and 20.0% Na.sub.2 O, can be used with advantage as a glass frit which is mixed with the radioactive mixtures to form the melt.
A typical aqueous radioactive waste which is incorporated into a borosilicate glass matrix is the highly active nitric acid containing waste solution (HAW) which is obtained during reprocessing of irradiated nuclear fuel and/or breeder materials after the common extraction of uranium and plutonium in the first cycle of an extraction process. A concentrate (1 WW) is obtained by evaporation and simultaneous partial decomposition of the excess HAW solution, and, if this 1 WW concentrate is to be solidified after intermediate storage, it is necessary to initially practically completely denitrate it, preferably with formic acid.
According to a process of W. Guber et al, as described in "Symposium on the Management of Radioactive Wastes From Fuel Reprocessing"; Proceedings of a Symposium organized jointly by the OECD Nuclear Energy Agency and the International Atomic Energy Agency, OECD, Paris, Nov. 27th to Dec. 1st, 1972, Organization for Economic Cooperation and Development, Paris, Mar., 1973, pages 489 to 512, the denitration with formic acid is effected continuously or in batches in a separate denitrator.
The free nitric acid and the nitrates of the transition metals are destroyed in this denitration process. Thus, with a pH of about 2, most of the transition elements are present in the denitrated 1 WW concentrate as difficultly soluble oxides, hydroxides, formiates, etc., and the noble metals are present in elemental form.
Gaseous reaction products are formed during the denitration process, and these gaseous products include CO.sub.2, N.sub.2 O and traces of N.sub.2 and NO. It is the aim of the denitration process to reduce corrosion by nitrous gases and their secondary products and not to charge the waste gases with nitrous gases. A further aim of the denitration process is to drastically reduce the ruthenium volatility of the easily volatile RuO.sub.4 produced in the oxidizing environment during the subsequent high temperature stages. The denitrated 1 WW solution is dried in a separate spray calcinator and is substantially calcinated, separated in a likewise separate filter tower, and transferred to the melting stage. The resulting calcinate is mixed with measured quantities of solid glass components, i.e., a mixture of glass forming substances or a prefabricated granulated basic glass, respectively, and is melted in a melting crucible. Depending on the fill level in the crucible, its discharge opening, which is closed by a glass plug, is melted open from time to time, so that the glass melt can be transferred to a chill mold.
The waste gases from the spray calcinator are cleaned a first time over sinter metal filter cartridges or candles and are freed of solids, the total decontamination factor being about 10.sup.4.
This previously-reported procedure of W. Guber et al has a number of drawbacks. The process is complicated and expensive with respect to time and personnel. Seen purely theoretically, a denitrator explosion cannot be completely excluded. Such a highly unlikely accident could occur theoretically if, for example, the reaction were stopped, but the feeder solution would continue to be measured in and the heating system would simultaneously malfunction and then, with uncontrolled return of the heat, an explosion-like exothermal reaction would start.
S. Drobnik has examined the possibility of performing the steps of denitration, spray drying and calcination continuously in one process stage, as reported at pages 37 to 40 of "Jahresbericht 1970 -- Abteilung Dekontaminationsbetriebe; Bericht der Gesellschaft fur Kernforshung mbH" (in translation, Annual Report for 1970 -- Department of Decontamination Operations; Report of the Gesellschaft fur Kernforschung m.b.H.), Karlsruhe No. KFK-1500 (June, 1972). For this purpose, an electrically heatable stainless steel pipe of 3 m in height and 70 mm diameter and equipped at its upper end with a spray nozzle was used to carry a simulated inactive, nitric acid fission product solution and formic acid which were fed in through the nozzle. Helium was introduced as the driving gas in order to facilitate the subsequent gas chromatographic examination of the waste gases. After passage through the spray dryer, the dried product was separated in a cyclone and the vapors were condensed in a cooler. The apparatus employed for these experiments was of the type with small throughput (laboratory equipment). In each experiment, 250 l of a model solution, which was 5.2 molar for hydrogen ions and about 7.1 molar for nitrate ions, and a 98% formic acid with a mole ratio of HCOOH: H.sup.+ of 2.55 were measured at a speed of about 5 ml per minute into the spray chamber which had been heated to 500.degree. C. The throughput for helium was 18 l/h. The dried product reached a temperature of 220.degree. to 300.degree. C. It was found that the reaction of the formic acid with the nitric acid and part of the nitrates takes place in the upper portion of the apparatus. In the lower portion, the remaining nitrates decompose to oxides and nitrous gases which themselves are reduced to N.sub.2, N.sub.2 O and NO by excess formic acid. Volatilization of ruthenium could never be proved.
This previously-reported process of S. Drobnik is also complicated and time consuming. The stainless steel pipe which is heated externally permits only a limited throughput of waste solution.